ML19309D562
| ML19309D562 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/08/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19309D561 | List: |
| References | |
| ACRS-1684, NUDOCS 8004100512 | |
| Download: ML19309D562 (22) | |
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.i MINUTES OF THE NOVEMBER 5,1979
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///S' ACRS AD H0C SUBCOMMITTEE MEETING ON THREE MILE ISLAND 2 ACCIDENT IMPLICATIONS WASHINGTON, DC The ACRS Ad Hoc Subcommittee on the Three Mile Island 2 Accident Implications regarding nuclear power plant design held an open meeting on November 5,1979 in Room 1046, 1717 H St., NW, Washington, DC. The purpose of this meeting was to discuss the implications of the March 28, 1979 incident at the Three Mile Island Unit 2 station on nuclear station design, policy, and criteria.
Notice of this meeting was published in the Federal Register on October 19, 1979. A copy of this notice is included as Attachment A.
A list of attendees for this meeting is included as Attachment B, and a schedule for this meeting is included as Attachment C.
Selected portions of the meeting handouts are included as Attachment D.
A complete set of handouts has been included in the ACRS Files. There were no written statements or requests for time to make oral statements received from members of the public. The Designated Federal Employee for this meeting was Mr. R. Major.
IMPLEMENTATION OF SHORT-TERM LESSONS LEARNED - T. Telford, NRC Staff Mr. Telford of the NRC's Division of Operating Reactors (DOR) gave a status report on the implementation of the short-term lessons learned recommendations. He noted the Staff's schedule remains fim, however, an October 30, 1979 letter to all operating nuclear power plants discusses the short-term requirements and reflects some additional flexibility over the original descriptions.
Mr. Telford described the TMI implementation group working on the short-term recommendations.
It is composed of a steering committee and four vendor ori-ented teams. Each team is to be composed of a multi-disciplinary group.
Mr. Telford explained that a September 13, 1979 letter to all licensees in-cluded additions to NUREG-0578. These additions were item recommended by the ACRS and endorsed by the Commission. These items include requirements for a containment pressure monitor, containment water level monitor, and containment H c ncentration monitoring system.
In addition, the Director 2
of NRR, through additional discussions and studies, has added the reactor coolant system venting requirement. NUREG-0578 minus certain items and with the addition of the four new items compose a set of features the Division of Operating Reactors is implementing on all operating reactors.
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.,h THI-2 Accident Implications 11/05/79 The October 30, 1979 letter to all licensees restates the position that was in either the NUREG-0578 report or the September 13,1979 D0R letter. The October 30, 1979 letter provides additional clarification for most of the items. The schedule for implementing the short-tem recommendations has not changed since the September 13th letter. Licensees are being asked to respond within 15 days of receiving the October 30th letter regarding any requirements for clarification and justification of any differences.
Mr. Telford presented a sampling of difficulties licensees are experiencing in meeting the Staff's schedule. The diversity requirement for containment isolation is a difficult problem for a few licensees to solve in the time frame requested. The issue of integrity of systems outside of containment i
required to handle radioactive material in the event of an accident is another area that is difficult in tems of schedule. Shielding reviews may have been inadequate in the past. Many licensees are attempting to re-review the shielding requirements of their plant. There are difficulties in meeting the single failure criterion for automatic initiation of auxil-iary feedwater systems.
Sampling installations at many facilities are inadequate and will require purchasing of equipment and new construction. There are equipment avail-ability problems in the area of high range monitoring instruments. There is a personnel availability problem concerning the shift technical advisor recommendation. There is some difficulty in meeting an onsite technical support center in the short-vs.'long-term while a pemanent facility is under construction.
Finally, there are problems in equipment availability for direct indication of the PORV.
Dr. Okrent questioned Dr. Mattson on efforts underway in Germany to give attention to various signals that could affect closure of the PORV block valve. Dr. Mattson was aware of the effort, but was unsure of the amount of attention the NRC Staff was giving to the German studies.
(Circuitry is being studied to automatically close the PORY block valve under a variety of indications that the PORV might be stuck open.)
Dr. Mattson agreed to make assignments to study this approach.
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TMI-2 Accident Implications 11/05/79 BRIEFING ON STAFF'S REACTION TO THE REPORT BY THE PRESIDENT'S COMMISSION ON TMI Dr. Mattson summarized the reaction during the previous week of the Office of Nuclear Reactor Regulation to the Kemeny Commission report. Those items which 1
address organizational recommendations in the Kemeny Commission report were noted as being outside the purview of the Office of Nuclear Reactor Regula-tion in the sense of having responsibility to initiate action. NRR reviewed the 44 major recomendations of the Kemeny Commission, and found that 33 fall within their purview. Those recommendations which are not considered within
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their scope of responsibility are: Al, 2, 3; B6; El, 2; F3, 5; and G1, 2,3.
Of the 33 items within NRR's purview, action has been initiated on 31.
In some cases, the action goes beyond that suggested in the Kemeny Report, in other cases, the action falls short of that recommended by the Kemeny Com-mission.
In the group of 33 recommendations applicable to NRR, three of the recom-mendations (A-6,A-8, and A-10) relate to the CP and OL process. Thirty items are applicable to operating reactors of these two items (E-5 and G-5) no action has been taken or recommendation made by NRR. Twenty-eiott of the recomendations are receiving varying degree!! of attention at operating power plants. Dr. Mattson cautioned again that some of the present actions fall short of the Kemeny recommendation. He noted that the review he was reporting was conducted within the first two days after receipt of the report. The major shortcoming of the NRC's actions to date, appears to be in the area of assurance of management and technical capabilities of the individual utili ties. The present action with regards to management and technical capabilities does not seem to be as decisive a program as appears to be warranted by the recommendations of the Kemeny Comission.
j The proposed correction to this shortcoming is an approach which would require each licensee to have an augmented on-site technical staff for operational safety surveillance. This may become an addition to the shift technical advisory and on-site operating experience evaluation groups required by the short-term lessons learned. A separate line organi-zation in each operating plant is envisioned, reporting independently of the power production staff. to high corporate management.
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TMI-2 Accident Implications 11/05/79 Dr. Mattson noted that the Staff was in substantial disagreement with one of the Kemeny Commission findings (G-12) which states "With the present organization, staff, and attitudes, the NRC is unable to fulfill its respon-sibility for providing are acceptable level of safety for nuclear power pl ants."
If the NRC were to accept that finding, they would have to shut down the operating plants. The Kemeny Commission itself made no such recommendation.
Dr. Mattson noted that the intent of the NRC will be take all the recomendations
- of the Kemeny Comission, and all of the recomendations of the various lessons learned efforts and develop an action plan for the implementation of those recommendations, first and with highest priority on operating reactors.
It is the Staff's belief that this effort takes precedence over any ongoing licensing activities. The intent is to put the recomendations side-by-side in a systematic synthesizing process, where goals in each subject aree addressed by these various recomendations could be derived, and specific tasks identi-fled for articulating the licensing criteria required to achieve those goals.
New licensing criteria developed would be backfitted on the operating plants.
Following application of criteria developed to operating plants, the Staff would turn to the application of criteria to pending OL applications.
In this process, the Staff will be required to decide which of the myriad of recommendations would need to be implemented prior to start-up of the pending applications. Dr. Mattson noted that the actions he outlined were being referred to as a licensing " pause".
Dr. Mattson commented that he thought that those people at national labs which are under contract to NRR, performing non-THI-related reviews would continue to do so for the next several months. There would be some manage-ment oversight in a caretaker-type role over the next few months.
Dr. Okrent explained that one of the functions of the Subcomittee meeting was to develop infomation so that the full Committee could: consider the possibility of zero power or low power operation of Sequoyah and decide if there would be further questions with regard to the proposed operation of Diablo Canyon.
Dr. Mattson informed Dr. Okrent that the Staff expects to take no action regarding either of those two matters in the next few months.
Dr. Mattson did request, however, that the Subcommittee entertain briefings L
TMI-2 Accident Implications 11/05/79 by the Staff on work done to date on Sequoyah and Diablo Canyon. He also noted some information on Salem 2 and North Anna 2 would be presented.
Dr. Okrent questioned Dr. Mattson on how it would be decided the pause could end.
Dr. Mattsonsaid that this issue had not been completely de-cided yet. What will be done in forming the action plan is to synthesize the various recommendations according to subject area. Then statements of purpose or goals within each of those areas would outline an action plan for developing criteria to implement and achieve those goals.
It is then expected that NRR will try to prioritize the goals to be achieved.
Following approval of the action plan by the Commission, it is hoped to implement the action plan on operating reactors. When this stage is reached it is expected there will be a considerable reassignment of Staff resources. A time frame of somewhere beyond January 1980 was mentioned.
It was also noted there would be high degree Congressional oversight and need for Congressional approval of the plan.
Dr. Mattson said that as of now, the length of the licensing pause is un-certain. There are basic policy questions regarding what to require and in what time frame, for operating plants. Then a time frame needs to be de-cided for pending OLs.
Dr. Okrent inquired about the NRR thinking for possible design changes for new cps. Dr. Mattson replied that the action plan may lead to requirements for cps and in plants under construction that are different than the require-ments for operating plants. Dr. Mattson also noted the recent decision by NRR to solicit information from B&W reactors in construction that is sufficieent to make a decision about whether or not to stop construction on those reactors.
A risk and reliability evaluation will be done in connection with answering this question. A method similar to the reactor safety study will be done for B&W reactors; the first plant to be studied will be Crystal Reiver. This work is being lead by the Probabilistic Assessment Staff in the Office of Nuclear Regulatory Research, with considerable involvement by the technical reviewers within NRR. This will be part of what Research has called their IREP, Integrated Reliability Evaluation Program.
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TMI-2 Accident Implications 11/05/79 Dr. Okrent inquired about reports regarding reactor sites located near high population density locations being shut down. Dr. Mattson noted he was not familiar with any such suggestions. He thought the siting policy task force and emergency preparedness task force might be more familiar with such ideas. He was not able to say whether shutdown had been given serious practical consideration.
LESSONS LEARNED FINAL REPORT - R. Mattson, NRR Dr. Mattson noted that the Lessons Learned Finr1 Report deals with more funda-mental basic issues than the short-term recommendations addressed. The final report addresses the three areas of nuclear power plant design, operation, and regulation.
It is structured to give prominence to the policy issues in those three areas. An appendix is included in the report that keys the major sections of the report to a set of specific recommendations aimed at bringing about change in those areas.
Dr. Mattson remarked that he had been advised that the short-term actions related to answering the short-term recommendation nearly saturated the technical expertise cf the Staff and the utilities themselves. He expected this is to be the case until after the January 1,1980 time frame for com-pleting items in the Short-Term Lessons Learned. Dr. Mattson said he did a superficial estimate of the resources required for the recommendations in the final lessons learned report. Two thousand man-years of technical work over the next two years would be required by industry to implement the "inal recommendations; it would require over 100 man-years for the NRC Staff.
Dr. Mattson said the fundamental lesson, in the view of the Task Force, from Three Mile Island is an operational safety lesson. Operations safety deserves paramount attention for the future. He noted only the joining of safe design and safe operation produces a safe nuclear power plant. Operations safety was defined as including human factors engineering,' personnel quali-fications and training, quality assurance of operations, and technical and management support of operations.
n TMI-2 Accident Implications 11/05/79 Dr. Mattson commented on the goals of operations safety. They are to reduce challenges to safety and to mitigate challenges to safety. By
" reducing the challenges" the task force means the challenges caused by human error. By " mitigating the challenges" the task force means maxi-mizing the capability of human beings to intercede productively in off-normal conditions.
In addition, reducing challenges to systems includes the recognition of personal responsibility and accountability at all levels of the operations organization.
Dr. Mattson discussed the topic of operations safety and who has responsi-bility for it. He noted the day-to-day hands-on attention to details only comes from the utility. However, NRC must set criteria and vigorously enforce them.
The last chapter in the Lessons Learned Final Report comments on the need for improvements in reactor regulation.
First and foremost the Task Force addressed the need to better articulate the reactor safety goal. Closely connected with a clearer definition of the safety goal is a better state-ment of the backfit policy.
Dr. Mattson called attention to a fundamental policy change on the part of the Staff in the final report having to do with the mitigation of beyond-design basis accidents. The final report stongly encourages design features for degraded core accidents and core melt accidents. This is a shift fran the prevention strategy of the past, and adds to a mitigation strategy for core melt accidents.
Dr. Mattson suggested reform of the licensing process to improve quality.
A system-level review was noted. Such.a review would audit a design, if a violation of criteria were to be found the design would be returned to the utility to redesign and perform a verification and validation check on the design to assure the design criteria had been met.
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TMI-2 Accident Implications 11/05/79 Attention to unresolved safety issues and implementing lessons learned from operating experience are tied together, Dr. Mattson noted.
In both cases it is necessary to dedicate the resources necessary to accomplish these goals and set end dates for completing action on these topics. Dr. Mattson also noted the NRR Staff will need to do some preparations of its own in order to supply technical support for emergency response.
Improvements in plant design were discussed. There were two significant areas that required further attention, that are closely related to the way the Staff presently does business. The first is dealing with how non-safety-grade equipment, as it has been called in the past, relates to safety, and the effects that non-safety grade equipment can have in causing and re-sponding to off-normal situations. A second area is the overylaying of deterministic requirements with reliability assessments. Dr. Mattson discussed the Integrated Reliability Evaluation Program (IREP). Particular designs would be examined for contributors to reliability or contributors to risk for all systems in a few lead plants. As the IREP goes through de-sign groups and as generic problems may be identified, program outlines might be extended at intermediate stopping points in the IREP to all operating plants.
Dr. Theofanous offered the suggestion that more attention be given to how a system responds under a variety of conditions and perturbations. He felt more attention needs to be paid to the mechanistic behavior of reactor response.
Dr. Okrent questioned the Staff on IREP. He asked who would conduct the study and what would be the time frame? The Staff replied that the. program would be conducted by the Staff and their contractors. At least a year and a half (an optimistic estimate) would be required to perform the program.
Dr. Okrent wondered why industry wasn't asked to perform IREP concurrently l
with NRC. The Staff replied because of the existing state of the technology results generated from different sources may produce results which are not comparable and interpretable on a common basis. Dr. Okrent urged considera-tion be given to allow licensees to perform IREP-type studies concurrently I
with the NRC Staff
TMI-2 Accident Implications 11/05/79 Dr. Mattson discussed the inclusion of design features for mitigation of degraded core and core melt accidents. The Task Force came to the con-clusion such design features should be required from two points of views:
one was how close TMI came to larger consequences and secondly from the three echelons of the defense-in-depth concept, where the third level is the consequence mitigation level embodied today in the containment leak rate -- siting philosophy. A topic which receives consideration in this area is the decay heat removal capability, using existing decay heat removal systems when the core is in a degraded state:(as an example, the existence of debris which occurred at Three Mile Island). Also included in this area would be equipment to cope with beyond-design-basis accidents such as equip-emnt associated with hydrogen control.
However, it was mentioned that any change in the hydrogen ' design requirements would require rulemaking.
In response to a question from Dr. Okrent, Dr. Mattson said that the Task Force had not received any comments from Mr. Denton, Director of NRR, on the Lessons Learned Final Report. They would expect his coments following those from the Committee, when NRR develops its action plan.
Dr. Okrent asked if the manning of control rooms at multi-unit sites had been explored, given a major. release from one unit to ensure the other units could remain manned and kept in a safe shutdown condition. He was informed the Task Force had not explicitly explored. this question.
It needs to be addressed as does the question of dealing with debris in RHR lines.
Dr. Mattson again noted that NRR will list all the actions required on the recommendations which have been made. The items will be categorized, put together by subcategories, and definitive action plans developed on all
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unresolved issues.
If items that are felt should receive attention are not identified in the plan, they should be called to the Staff's attention.
He noted it will be difficult to make any comment on the sufficienty of the Staff's response to TMI issues until after the action plan is developed.
TMI-2 Accident Implications 11/05/79 Dr. Catton inquired about the advisability of a dedicated high pressure decay heat removal system. The Staff noted that the current view of such a system is that it is unnecessary if the methods of operation and training associated with current systems is improved as is the current case.
STATUS OF THE SEQUOYAH LICENSING REVIEW Mr. Shao summarized the status of the non-TMI review items. A list of these items is included as page 1 of Attachment D.
There are a total of 40 items, 24 of which were identified in the March 1979 Safety Evaluation Report. Action on 29 of the 40 issues is currently completed and action on the remaining 11 issues is expected to be completed in about a month. The items on which action is completed are indicated as checked items on page 1 of Attachment D.
Mr.
William Cottle, the resident inspector at the Sequoyah plant, gave the I&E status report for the Sequoyah plant.
It is noted that construction of Unit 1 is essentially complete. The only area in the preoperational test program j
which is not yet complete is physical safeguards.
I&E efforts are currently f
concentrated on the follow-up on the I&E Bulletins responses and follow-up on construction deficiency report items. He indicated that he expected all of these items to be closed-out within three to four weeks. Dr. Okrent indi-cated that the Committee may want to look at the implementation of the I&E Lessons Learned Report on the Sequoyah plant before the issuance of the full power operating license.
Mr. Robert Stratterfield discussed the possible effects of a high energy pipe break on the calibration of steam generator level measurement systems. This could, in some cases, lead to a delay in the system actuation and could have effect on the post-accident monitoring. The short term fix is to adjust set-points to alleviate this problem. Mr. Stratterfield also discussed the issues involving possible control system failures during high energy pipe breaks.
Westinghouse, in the course of the review of what equipment might be affected by such a pipe break identified four systems which could, if they were to mal-function, affect the outcome of accident events. The systems were the steam generator PORV control system, the pressurizer PORV control system, the feed-water control system, and the automatic rod control systems. Westinghouse will be performing a plant by plant analysis of these events.
TMI-2 Accident Implications 11/05/79 Dr. Okrent noted that the NRC Staff in the ASLB bearings on St. Lucie had been asked by the Board to address why the sequence involving the loss of offsite power combined with the failures of all diesels to start should not be taken into consideration in the design of the plant and asked if this was to be taken into consideration for the Sequoyah plant. The Staff noted that this event (the loss of AC power) was currently an unresolved safety item. Dr. 0krant indicated that he believed that this event should be further discussed with regard to Sequoyah.
The implications of issuance of low power licenses was discussed. Mr. Vassallo j
noted that the question was covered by Mr. Mattson in the morning's discussion.
Mr. Mattson indicated that first priority for the allocation for NRC resources would be given to establishing licensing criteria and requirements for the Kemeny recomendations, and after that resources would be allocated to the implementation of the licensing criteria / requirements in operating plants.
Priority for allocating resources to operating licenses reviews would come after operating plant reveiws. Mr. Mattson stated that in his opinion no action would be taken on operating licensees within the next few months.
Mr. Knight summarized the status of the Sequoyah seismic evaluation. Site dependent response spectra have been developed and discussed with the ACRS at previous meetings. The Staff has verified the competence of the Sequoyah plant relative to the new site dependent spectra on an audit basis. Mr. Anight estimated that the audit performed by the NRC encompassed about 301 of the structures and equipment necessary for safe shutdown and about 10-20% of other plant equipment, and that they believed that the items sampled were represen-tative.
Mr. John Long discussed the status of the Staff review of UHl. Mr. Long indi-cated that the Staff has been concerned about the possibility of nitrogen injection into the primary system during water injection from the upper head accumulators and have taken measures to minimize the probability of this occurrence. Series valves are used to isolate the upper head accumulators after the water injection. Only a minimal amount of nitrogen is dissolved in the upper head iniection water as the water is separated from the nitrogen
TMI-2 Accident Implications 11/05/79 gas by ruptrue disc prior to injection and water is maintained in the tank prior to the shutoff of the upper head injection to preclude gas ingestion by vortexing. Semiscale tests are planned to assess the effect of' failure of the series isolation valves to close.
Mr. Michael Williams discussed the status of the NRR review of Sequoyah's responses to the Lessons Learned recomendations. He indicated that responses had been received from TVA on a timely basis and that NRC was currently review-ing them. He indicated that he believed the review would be completed by January.
Dr. Okrent and Dr. Mark discussed whether or not the Subcommittee should recommend that the Sequoyah application be brought before the Full Committee as scheduled at the November meeting.
It was their opinion that in view of the NRR's intention to institute a licensing pause that it would not be a worthwhile expenditure of the Comittee's time to review this application at this time.
It was decided that the Subcomittee would report to the full Committee on what had transpired at this Subcomittee meeting.
NRC RESPONSE TO ACRS QUESTIONS REGARDING NEAR TERM OL PLANTS Mr. William Milstead addressed the questions dealing with the Sequoyah and McGuire plants. Mr. Williams noted that the Staff had reviewed the Sequoyah and McGuire plant design from the standpoint of TMI-2 accident implications and had concluded that the only plant features which were very unique to these plants are the use of the ice condenser containment and the upper head injection systems. Mr. Williams noted that the ice condenser was a small volume, low design pressure containment, and that at this point in time the Staff had concluded that the structural integrity of the containment building could not be assured if hydrogen, in the amount generated in the TMI-2 accident, were to be released and burned in a ice condenser containment building. The present ice condenser design contains features which are intended to prevent an uncon-trolled hudrogen burn. There are redundant safety grade recombiners (100 CFM W thermal recombiner units) and redundant trains of mixing systems (the return air system which is used to recirculate air between the upper and lower contain-ment compartments and a hydrogen skimmer system is used to prevent localized accumulations of hydrogen in containment subcompartments). The design basis 1
i TMI-2 Accident Implications 11/05/79 for these systems was a 5% metal-water reaction.
It would appear that these ti systems, as designed, could accommodate up to about 15% metal-water reaction.
Mr. Milstead noted that the recommendation of the Lessons Learned Task Force was to resolve the question of hydrogen control in rulemaking. He additionally noted that design basis would have to be such as to accommodate gases that would be released through the high point vent valves which are required by the Lewsons Learned recommer.dations.
Non-condensable gas accumulation in the steam generator U-tubes was discussed.
It will have to be established that even with non-condensable accumulation the decay heat can be removed.
Mr. Milstead indicated that applicants would be required to perform a small break analysis using both the UHI ECCS evaluation model combination with the best estimate input as is being done by non-UHI plants. The semiscale test program includes a series of tests which is intended to investigate the effect of the upper head injection.
The use of filtered vented containment systems for ice condenser plants was briefly discussed. Mr. Butler indic;ted that it was the NRC's intent to evaluate this matter and that the Research Office of NRC has contract work at the Sandia Laboratories which is directed towards exploring these concepts.
Mr. Rowsome discussed means of controlling large quantities of hydrogen in ice condenser containments. He noted that at the October 3rd TMI Implications meeting he had been asked to prepare a preliminary recommendation as to by what means hydrogen control might be achieved. He indicated that he would presently favor distributed ignition sources but that this recomendation was of a preliminary nature. He indicated that it was his belief that either inerting or distributed ignition sources by themselves would be ineffective in altering the consequences of core melt accidents. The predicted failure in the core melt accident is by containment overpressure and even, if hydrogen burnes were l
prevented, the failure would occur later in the scenario by other means.
It was noted that a hydrogen control would be effective in preventing containment overpressure for accidents less than a core melt.
It was noted that on the basis
TMI-2 Accident Implications 11/05/79 of the present knowledge there would be a favorable value impact for the filtered system on an ice condenser containment.
_ BRIEFING ON LICENSING PAUSE - H. Denton, Director, NRR Mr. Denton made a brief statement to the Subcommittee on the licensing pause.
Before proceeding with licensing of any plants that are presently under review, the Commission will await the Staff's assimulation of the short-term lessons and the long-term lessons, the President's Comission recomendations, the rulemaking on emergency planning, some decision on siting, and the Rogovin Comission's suggestion for a new licensing framework. He believed the j
earliest such an action plan could be devised by the Staff would be next spring. The Comission would then proceed on a case-by-case basis in the plant-by-plant confomance to the results of their assimulation of all of these requirements.
In response to a question from Dr. Okrent concerning licensing reactors at power levels less than full power until decisions could be made regarding the implications of the TMI recomendations, Mr. Denton responded that presently he felt such proposals would not be well received.
In response to a question from Dr. Mark, Mr. Denton expressed the opinion that next spring is the earliest he coulf foresee the licersing pause ending.
BREIF STATUS REPORT ON EMERGENCY PR$CEDURES Mr. Williams of NRR gave a brief status report on the Staf f's review of emergency procedures. Presently the Staff is looking at emergency a perating procedures on near-term OLs including Sequoyah, Diablo Canyon, North Anna 2, and Salem 2.
He noted as a result of the short-tenn lessons learned re:ommendation 2.1.9, transient and accident analysis, the effort includes reanalysis of the small
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break LOCA, feedwater transients and inadequate core cooling, and then doing all chapter 15 transients and accidents again. Each reavalysis will involve separate guidelines for the vendors and an incorporation of these guidelines in terms of what is learned during the analysis into a revision of the plants and procedures. These procedures will all be reviewed by the Staff.
TMI-2 Accident Implications 11/05/79 Mr. Williams also noted that there is a recommendation in the long-term lessons learned that the Staff get involved in an interdisciplinary team effort that includes such people as experts in crisis management, in theories of education, in terms of taking a look at operating procedures. Currently, NRR has requested copies of the emergency procedures for the near-term OLs.
Selected procedures will be audited by the review team headed by Mr. Williams.
Current plans also include a discussion between applicant, vendor, and Staff to detemine their interrelationship in terms of their preparation and their review and approval of emergency procedures. The Staff will also try to determine within individual organizations the chain of comand in terms of review responsibility for emergency procedures.
It is also anticipated to observe a procedure performed on a simulator. Another possibility is to go to each plant and walk through an emergency procedure at the plant.
Follow-ing the audit of emergency procedures, a report will be issued containing the Staff's recommendations.
RELIABILITY OF RESIDUAL HEAT REMOVAL SYSTEMS - T. Speis Work is underway in this area by the Staff. Dr. Mattson, in response to a question from the Subcommittee, said that in the Staff's judgment a dedi-
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cated high pressure shutdown cooling system was not a potentially productive way to address the questions raised by TMI. He noted such a system adds to consequence prevention capabilities; the Staff's thinking is that a point of diminishing returns has been reached with prevention capability and what would be productive in a move toward mitigation capability.
STATUS OF TMI-RELATED SEISMIC REVIEW AT DIABLO CANYON - J. Knight, NRR-DSS Mr. Knight of the Staff described an interdisciplinary review of Diablo Canyon to address the expected performance of electrical and mechanical systems and equipment that has not been seismically qualified and based on that expected perfomance assess the assurance that cold shutdown can be accomplished and maintained, given a severe earthquake up to and including the SSE. This review is aimed at resolving concerns raised by the Subcomrr.ittee concerning
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TMI-2 Accident Implications 11/05/79 I
TMI-2 implications at Diablo Canyon. Mr. Hoch of PG&E agreed such a program would be useful and noted he would initiate such a review at Diablo Canyon with or ' ithout NRC Staff participation.
w Dr. Mattson also noted that other near-term OLs may be required to perform studies on the influence of failures in the secondary or non-safety-grade I
equipment on the rest of the plant.
PRESSURIZER HEATERS - Potential Failure Modes for Pressurizer Heaters and Associated Consequences Dr. Mattson replied to a question raised during the October 3,1979 meeting of the TMI-2 Accident Implications Subcommittee regarding the potential for pressurizer pressure boundary failure if the pressurizer heaters should be-come uncovered and overheat. The failure mode most important to primary system integrity, identified, was an event which caused a pinhole in the sheathing of the pressurizer heaters. Once water gets by the sheathing it can run into the Calrod heaters. A concern can be raised as to whether reactor coolant could escape through the hule, out the heater wire pene-trations, and into containment resulting in a small LOCA. This does not appear likely since for some designs heater penetrations are designed and tested to prevent leakage. For other heaters, there is a seal area six inches long. A leak through this area would resuli: in a very small amount of steam leakage to containment, too small to be classified a small LOCA.
Additional information on the Staff's investigation into the this area is contained in a memo from R. Mattson to R. Fraley dated November 5,1979 (attached). The Staff also agreed to continue their investigation of this question.
WATER SUPPLIES'FOR AUXILIARY FEEDWATER SYSTEMS Dr. Okrent asked the Staff if it was true that some operating reactors do not have Class I water supplies for the auxiliary feedwater system. He was j
infonned this was correct. Dr. Okrent inquired about how the question of
t TMI-2 Accident Implications 11/15/79 suitability of this situation was to be resolved. The Staff replied that studies on auxiliary feedwater reliability would be handled by the Bulletins and Orders Task Force. Dr. Okrent requested documentation when available that would describe the technical basis for resolving this item.
DERATING OF PLANT POWER Dr. Okrent inquired if there was any relative increase in safety if a plant is run at some lower percentage of full power.
In general, the Staff replied that intermediate power levels do not greatly improve safety. This is especially true in the area of performance of operators, multiple failures, and beyond-design basis events.
However, this was a preliminary comment on the part of the Staff and they are still keeping an open mind on the topic.
NOTE:
For additional details, a complete transcript of the meeting is available in the NRC Public Document Room,1717 H St., NW, Washington, DC 20555, or from Ace-Federal Reporters, Inc., 444 North Capitol St.,
NW, Washington, DC 20001.
30448 Federal Register / Vol. 44. No. 204 / Friday. October 19. 1979 / Notices i
provisions of the Act and the Code.
NATIONALCOtHMSSION ON AIR ne agenda for subject meeting shall i
mcluding statutory or administratis e QUALITY be as follows:
j esemptions and transitional rules.
Monday. November 5.1979,838 am.-
Furthennore, the fact that a transaction Meeting Scheduled for November 13 until the conclusion of business is subject to en administrative or De National Commission on Air The Subcommittee may meet in statutory exemption is not dispositive of Quality hereby gives notice of a meeting Executive Session. with any ofits whether the transaction is in feet a scheduled for November 13.ne meetmF consultants who may be present, to l
prohibited transaction.
will be held in Room 4200 of the Dirksen.
explore and exchange their preliminary Senate Office Building. located at First opinions regardmg matters which should Written Comments and Hearing Street. NL and Constitution Avenue.
be considered during the meeting and to R'9"
NL Washington. D.C., and will begin at formulate a report and recommendation All interested persons are invited to 1:00 p.m.
to the full Committee.
submit wntten comments or requests for ne agenda for the meeting will At the conclusion of the Executive a hearing on the pending exemption to include the following items:
Session. the Subcommittee will hear the address above, within the time
- 1. Approval of the minutes of the presentations by and hold discussions penod set forth above. All comm-nts October 5.1979 Commission meeting.
with representatives of the NRC Staff, will be made a part of the record.
- 2. D:scussion of activities of the the nuclear industry.various utilities.
Comments and requests for a hesnna Commission's Research Committee.
and their consultants, and other should state the reasons for the wnter's
- 3. Consideration and selection of interested persons, regarding the interest in the pendmg exemption.
alternative air pollution control policies implications of the DC-2 Accident.
Comunents received will be available for. to be applied in the 'hi== ion's in addition,it may be necessary for public inspection with the application regional stud-a the Subcommittee to hold one or more Questions about th; 2eeting should closed sessions for the purpose of for exemption at the address eet forth be directed to Mr. Morris A. Ward at exploring matters involving proprietary above.
(202) 245-6355.
information. I have determined. la Proposed Exemption Nationel Comsnission on Air Quality.
accordance with Subsection 10(d) of Pub. L 92-463, that, should such Based on the facts and Williara H. tawin pr.,
sessions be required it is necessary to representations set forth in the Director.
cl se these sesalons to protect application, the Department is in on was fed m-in-m s es ami I
a information (5 U.S.C.
]h{'4 considering granting the requested swwo coes sem es-a exemption under the authority of sectior' Further information regarding topics 40e(s) of the Act and section 4975(c)(2) to be discussed, whether the meeting 1
of the Code and in accordance with the NUCt. EAR REGULATORY has been cancelled or rescheduled. the procedures set forth in the ERISA COMMISSION Chairman s ruling on tequests for the cedum 1
18 1A 28.
ty to resen oral statem ts Advisory Comn(ttee on Reactor
, opgo g
restrictions of sections 406(a). 406(b)(1).
tain d y a prepaid t lepho e call to U 2 e
and 406(b)(2) of the Act and the taxes the cognizant Designated Federal impuestions Me Nuctew Powwplant imposed by section 4575(a) and (b) of Employee. Mr. Richard K. Major Design; Meeting the Code, by reason of section (telephone 202/634-1414) between 6:15 4975(c)(1)[A) through (E) the Code shall The November 7,1979 meeting of the a.m. and 5 00 p.m EITI' before, and EST -
l not apply to the cash sale by the Plan of ACRS Ad Hoc Subcommittee on the after. October 28,1tr19.
I certain real property located at 1211 Three Mde Island. Unit 2 Accident-Background information concerning i
North loring Street in Portland. Oregon Implications Re Nuclear Power Plant items to be discussed at this meeting for $380.000 to Mr. William Anderson Design has been rescheduled to be held can be fo'und in documents on file and provided that this amount is at least the on November 5.1979 in Room 1066.1717,available for public inspection at the fair market value of the property.
H St NW Washington, DC mm NRC Public Document Room.1717 H
%e proposed exemption. if granted.
Notfce of this meeting was published Street. NW. Washington. DC 20555 and Octobe 18.197E at the Covernment Publications Section.
will be subject to the express conditions In accordance with the procedufes State Library of Pennsylvania, that the material facts and outlined in the Federal Register on Education Buildir.g. Commonwealth and representations contained in the application are true and complete, and Oc,tober 1.1979. (44 FR 5640s). ora Walnut Street. Harrisburg. PA 17126.
wntten statements may be present by that the application accurately describes members of the public, recordings will Dated: October ts 187s.
all material terms of the transaction to be permitted only during thou patio 6s Joha C. Hoyle, be consummated pursuant to the of the meeting when a transcript is being Advisory commisse Memgement officer, exemption.
kept, and questions may be asked only ps w emrawwwmees t Sipied at Washington. D.C., this 13th day by members of the Subcommittee,its ow,,o coos -
of October ters.
consultants, and Staff. Persons desarms Ian &ImmeN.
to make oral statements should notify the signated F plo as far A so Conuni Administator. h>nsion and Wetfore Benefit l
Programs. LaborsMonagement Services Pr*
A dminissrotson, Department aflo6or.
appropriate arrangements can be made Components; Joint Meeting to allow the necessary time during the ne ACRS Subcommittee on Metal pu o. u s-anusraw so. w m ees mi meeting for and stataan=ata.
Components will hold an open meeting
== i=== cosa es-me
l o
Feder:1 Register / Vcl. 44 No. 204 / Fridsy, Oct:ber 19, 1979 / N:tices 80449 on November 5,1979, in Room 1187,1717 IDocket No. 40-47451 Commission Division of Waste Mansgement,483-SS, Washington, D.C.
l H St., NW Washington, DC 20555.
Availability of Environmental Reht 20555, Phone (301) 427-4103.
In accordance with the procedures n tt outlinedin the Federal Register on Dated at Silver Spring. Maryland. this 9th
.ntal tS t
October 1,1979 (44 FR 56408), oral or d*F* * **
- Concoming Issusence ref a Source written statementa may be presented by Material Ucense for the Bison Basin Foy the U.S. Nuclest Regulatory Comminh members of the public, recordings will Project To Be Locatedin Fremont Roos A.3carano,
[
be permitted only during thosq portions County, Wyo.; Ogle Petroleum,Inc.
Chief. Umnium Recovery Ucensing Bmnch.
of the meeting when a transcript is being I
Aossocy:U.S. Nuclear Regulatory Division of waste Management.
I kept, and questions may be asked only Commission-p wwme neo.is.ws w by members of the Subcommittee,its consultents, and Staff. Persons desiring Actiom Notice of Intent to Prepare a same caos riewwe to make oral statements should notify Draft Envrionmental Impact Statement.
the Designated Federal Employee as far sunsasAny:1. Description of the Proposed Advisory Committee on Reactor in advance as practicable so that Action--Ogle Petroleum,Inc., proposes Safeguards, Subcommittee on the appropriate arrangements can be made to construct and operate an in-situ Sequoyah Nuclear Power Station; to allow the necessary time during the uranium extraction facility at the Bison Meeting meeting for such statements.
Basin Project site. This project is located De ACRS Subcommittee on the,
The agenda for subject meeting shall in southwestem Wyoming be as follows approximately 50 miles south of Sequoyah Nuclear Power Station will Riverton. and 30 miles southwest of hold a meeting on November 5,1979,in Room 1046,1717 H Street, N.W.,
Monday, November 5,1979,8:30 as ~
jefirey City.The 400.000 pounds of Washington, DC 20555 to review the until the conclusion of buelness uranium per year capacity in-situ extraction facility will be located in application of the Tennessee Valley The Subcommittee may meet in Sweetwater River Drainage Basin.
Authority (TVA) for a permit to operate Executive Session, with any oiits Section 25 T27N, R97W Fremont Units 1 and 2 of this station.
consultants who may be present, t in accordance with the procedures Coun. Wyomin explore and exchange their preliminary 2.T e principsfaltematives currently outlined in the Federal Register on opmions regarding matters which should planned to be considered include October 1,1979 (44 FR 56408), oral or be considered during the meeting and to altematives of siting, waste written statements may be presented by formulate a report and management methods, energy sources, members of the public, recordings will recommendations to the full Committee.
and the attemative of no licensing be permitted only during those portions At the conclusion o{the Executive action.
of the meeting when a transcript is being Session, the Subcommitee will hear 3.ne scoping process willinclude a kept, and questions may be asked only presentations by and hold discussions meeting to be held in the Lodge Room of by members of the Subcommittee,its 5
with representatives of the NRC Staff, the Elks,207 E. Main Street, Riverton, consultants, and Staff. Persons desiring the Boiling Water Reactors (BWR)
Wyoming, on November 1,1979, at 10 to make oral statements should notify Owners Group. and their consultants, a.m. This meeting will provide for a the Designated Federal Employee as far regarding the matter of BWR pipe briefing of interested parties concerning in advance as practicable so that cracks. ACRS generic items pertinent to the proposed action and alternatives appropriate arrangemhnts can be made the purview of this Subcommittee such and opportunity for comment on the to allow the necessary time during the as stres's corrosion cracking in BWR scope of the proposed statement.The.
meeting for such statements.
piping. and inservice inspection of participation of the public and all The agenda for subject meeting shall interested govemment agencies is be as follows: Monday, November 5.
reactor coclant pressure boundary will invited. Copies of this notice will be 1979,8 30 an until the conclusion of aIso be addressed.
mailed to all affected federal, state, and business.
Further mformation regarding topics local agencies, and other interested The Subcommittee may meet in to be discussed, whether the meeting persons. Written comments conceming Exec 6tive Session, with any ofits has been cancelled or rescheduled, the the scope of the proposed statement will-c4nsultants who may may be present, to Chairman's ruling on requests for the be accepted until November 9,1979.
explore and exchange their preliminary opportunity to present cral statements 4.The DEIS is expected to be opinions regarding matters which should and the time allotted therefore can be available to the public for review and be considered during the meeting and to j
obtained by a prepaid telephone call to comment in February,1980.
formulate a report and the Designated Federal Employee for 5.The applicant's Environmental recommendations to the full Committee.
this meeting. Mr. Elpidio G. Igne, Report and Appendix and any - -
At the conclusion of the Executive (telephone 202/834-3314) between 8:15 subsequent documents will be available Session. the Subcommittee will hear a m. and 5:00 p.m., EDT before, and EST for inspection and copying at the Fublic presentations by and hold discussions after October 28,1979.
Document Room (PDR).1717 H Street, with representatives of the NRC Staff, N.W., Washington, D.C. 20555. Copies of TVA, and their consultants, pertinent to Dated. October 15.1979 the Environmental Report and Appendix this review.
lohn C. lloyle are also being provided to the State in addition,it may be necessary for Adnsory Committee Management Officer.
Planning Coordinator Office of the the Subcommittee to hold one or more h
Governor,2320 Capitol Avenue, closed sessions for t. e purpose of p om uniz ra.4 so.is., e es..
Cheyenne, Wyoming 82002.
exploring matters involving proprietary sumo coca ysm,.m Questions about the proposed action, information.1 have determined,in DEIS, or scoping meeting and any accordance with Subsection 10(d) of written comments should be directed to Public Law 92-463, that, should such l
R. S. Kaufmann, U.S. Nuclear Regulatory sessions be required, it is necessary to i
fiO450 Federal Register / Vcl. 44. Ns. 204 / Pdd:y. October 19, 1979 / Notices r
OPM/INERNAlce Employee locator close these sessions to protect proposed its systems of records (44 FR Card File (44 m 30853, formedy OPM/
proprietary Information (5 U.S.C.
30836. May 29,1979) identified as OPM/
552b(c)(4)).
INTERNAIA through OPM/
LVI'ERNL15).
Further information regarding topics INTERNAIA2 OPM/CEN*mAI,1 OPM/LVERNAlr10 Employee through OPM/CE. EAL-14. and OPM/
Production Records (44 FR 30854, V
to be discussed. whether the meeting has been cancelled or rescheduled, the GOVT-1 through OPM/ GOVT-6, a formerly OPM/INERNAIAS).
Chairman's ruling en requests for the Report on New Systems was also filed OPM/ INTERNAL-11!nvestigator opportunity to present oral statements with OMB and Congress.%e required Performance Records (44 FR 30655, and the time allotted therefor can be 60-day advance notice period ended on formerly OPM/INERNAle17).
obtained by a prepaid telephone call to July 30.1979. Comments were received OPM/INTERNAlr12 Speaker Resume the Designated Federal Employee for concerning only some of the proposed and Clearance Records (44 FR 30856, this meeting. Dr. Richard P. Savio, systems, which are not being adopted by formerly OPM/LVERNAL-18).
OPM/INTERNAlr13 Motor Vehicle (telephone 202/634-32871 between 8.15 this notice, and are still under a m. and 5:00 p.m.. EDT (EST after consideration. Adoption of those Operator and Amident Report Records
~
October 28.1979).
systems will be announced at a later (44 FR 30857.formerly OPM/
INTERNAL-19).
Background information concerning date. While there were no commenta OPM/CENmAL-1 Civil Service items to be considered at this meeting received pertaining to the systems Retirement and Insurance Records (44 can be found in documents on file and adopted at this time, three notices do FR 30865. formerly OPM/ CENTRA!A).
i available for public inspection at the require non-substantive text changes The former CSC-4 Civil Service NRC Public Document Room.1717 H based on intra. agency coordination and Retirement Records system is Street. N.W Washington. DC 20555 and several systems have received new completely replaced by this system.
at the Chattanooga-Hamilton County.
numerical designations. Where no ON/CEN' MAL-6 Intergovernmental Bicentennial labrary.1001 Broad Street, changes to the text of the notice are Personnel Act Assignment Records (44 Chattanooga. TN 37402.
required, only the title, new numerical FR 30809. formerly OPM/ CENTRAL-7).
designation. and specific Federaj OPM/CENIRAle6 Administrative Dated. October 16. Ig-g.
Register page number are cited. Where Law Judge Application Records (44 m John C. Hoyle Advisory CoEmrtee Management Officer.
text, changes occur, the title and Federal 30671. formerly OPM/CENTRAle8). '
Register page number will be cited and OPM/CENTRAlr7 IJtigation and ira oauro r.w w.e m the text of the changed portion of the Claims Records (44 FR 30872, formerly
~* '""
notice will be shown. Where the ON/CENTRAleG).
adopted system completely replaces an OPM/CENTRAle6 Privacy Act/
existing Civil Service Commission Freeoom ofInformation Act (PA/FOIA)
OFFICE OF PERSONNEL system (in some cases the CSC system Case Records (44 FR 30674. formerly MANAGEMENT must remain in effect to cover records OPM/CENTRAle10).
Privacy Act of 1974; Adoption of-maintained by the Ment Systems OPM/CENTRAlee Personnel Protection Board) this is so noted. The -
Investigations Records (44 FR 30876.
Systems of Records systems adopted appear below.
formerly OPM/ CENTRAL-11).The Acacy: Office of Personnel Onion ofPer Anal Management.
former CSC/GOG4 Pusonnel Management.
Roderidt S. Speer.
Investigations Records system is AcT>oec Notice; Adoption of Systems of A#8I8##"# I##" "C' 878'*" M8"C8'#-
comple.tely replaced by this system.
RML ON/ CENTRAL-10 Directory of The following systems are adopted as Federal Executive Institute Alumni (44
SUMMARY
- The Office of Personnel published, with. in some cases, new FR 30878 formerly OPM/'INTRAL-12).
Management (OPM) has previously numerical designations as the only ON/CENTRAlett Presidential published a notice of several proposed change Management Intern Program Records (44 Privacy Act systems of records.The OPM/ INTERNAL-1 Defense FR 30lL79 formerly OPM/CENTRAI-13).
purposes of this notice are to:(1) 12 n mMEency Cadre Records, TheformerCSC-9 Presidential Identify certain of those proposed Management Intern Program Records systems of records, where no comments OPM/ INTERNAL-2 Negotiated
- system is completely replaud by this were received, as being adopted;(2)
Grievance Procedure Records (44 FR system.
a identify some non. substantive changes 30839. formerly OPM/IN'ERNAL-3).
OPM/CENIRAle12 Federal to those adopted; and (3) serve in part to meet the Privacy Act requirement for OPM/ INTERNA 1,3 Security Office Automated Career System (FACS) j annual publication of notices of systems Control Cards (44 FR 30640. formerly Records (44 FR 30880. formerly OPM/
j of records. This action is required by the OPM/INTERNAle4).
CENTRAle14).
creation of OPM by the President's OPM/INTERNAle4 Employee OPWINTERNAL-4 l
Reorganization Plan No. 2 of 1978, and Occupational Health Program Rece(ds to implement the Pnvacy Act, and has (44 FR 30841. formerly OPM/
System nome:
the effect of establishing systems of INTERNAle6).
Appeal and Administrative Review records for use by OPM and agencies.
OPM/INTERNAle6 Pay. Leave, and Records (44 FR 30844. formerly OPM/
EFFECTIVE D ATE: October 19.1979.
Travel Records (44 FR 30842 former!?
INTERNAlc7).
FOR FURTHER INFORMATioed CONTACT:
OPM/INTERNAle6).
l William H. Lynch. Work Force Records OPM/INTERNAlr7 Complaints and
!l Management Branch. Agency Inquiries Records (44 FR 30645, former13 cartoonies or asconos as tua sysme Compliance and Evaluation (202) 254-OPM/INERNAlr6).
This system contains records relating j
J 9778.
OPM/INTERNAle6 Employee to various appeal or administrative l
supetrMENTAny inconMATION:When Assistance Program Records (44 FR review procedures available to OPM the Office of Personnel Management 30851. formerly OPM/INERNAle13).
employees.These appeals or D**D 3 D l3~ Y f
~1 Mb
i ACRS AD H0C SUBCOMMITTEE MEETING ON THREE MILE ISLAND 2 ACCIDENT IMPLICATIONS NOVEMBER 5,1979 WASHINGTON, DC ACRS NRC STAFF D. Okrent, Chairman L. Rubenstein W. Mathis R. Denise C. Mark S. Varga C. Michelson, ACRS Consultant D. Vassallo I. Catton, ACRS Consultant R. Cudlin T. Theofanous, ACRS Consultant R. Mattson R. Major, Designated Federal Employee R. Satterfield R. Savio, ACRS Staff C. Stahle F. Litton P. Stoddart TENNESSEE VALLEY AUTHORITY J. Long H. Krug D. Ormsby O. Parr K. Whitt V. Moore L. Mills J. Conran D. Lambert W. Butler J. Ballentine R. Williams
- 0. Zeringue G. Holahan J. Bynum Z. Speis J. Carter R. Ireland D. Urkon J. Stolz J. Raulston R. Birkel A. Pagano J. Knight D. Morgan R. Tedesco C. McLaughlan J. Holman R. Guthrie PACIFIC GAS & ELECTRIC DUKE POWER COMPANY J. Hoch T. Heitman B. Norton G. Copp P. Crane J. Beall J. Schuyler M. Sills M. Furbush G. Sansbury G. Schnabil 0FFSHORE POWER SYSTEMS NORTHERN STATES POWER COMPANY R. Bruce P. Krumpos ILLIN0IS POWER COMPANY NORTHERN INDIANA POWER SERVICE COMPANY C. Wheeler P. Harrigan
%C. TOS
. WESTINGHOUSE ELECTRIC CORPORATION DAIRYLAND POWER G. Lang B. Angle J. Hord P. Docherty son ELECTRIC POWER RESEARCH INSTITUTE M. Siano R. Slahtkopp R. Sero R. Leyse M. Harding L. Conway GENERAL ACCOUNTING OFFICE POWER AUTHORITY FOR THE STATE OF NEW YORK S. Higgins R. Rajaram.
BBR COMMONWEALTH EDISON COMPANY K. Layer A. D. Rossin BROOKHAVE NATIONAL LABORATORY CAROLINA POWER & LIGHT COMPANY J. Weeks R. Hanford GPW SERVICE CORPORATION LOUISANA LOUISANA ELECTRIC LIGHT & POWER COMPANY F. S. Giacobbe K. Harrington BECHTEL CORPORATION EBASCO SERVICES INC.
B. Montgomery M. Pierson PUBLIC Fliesheler - CLIPS R. Heer.- Ace Federal Reporters, Inc.
M. Shiba - JAERI E. Hume - L.A. Times i
L-
l,
' -l..,, r.-.
.... -~.. - -.:. _ r
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(
ENTATIVE SCHEDULE FOR WE AD HOC SUBCOiMITTEE MEETING ON '!EREE MILE ISIAND 2 ACCIDDTP IMPLICAT i
AND l'
SEQUOYAH NUCLEAR STATION NOVDiBER 5,1979 3 hrs.
B:30 a.m.
REPORT FRO 4 'IEE LESSOIS IIARNED 'IASK FORCE (1) Long-Term IAssons Imarned (2)
Implementation of the Short-Term I4ssons 12:30a w (.0/
? Y
$N 4W W -
W 30 a.m.
DISCUSSIONS OF SEQU
/ ICE CONDENS
- UMI PLAN'IS 3 1/2 hrs.
(1) Status of NRC Staff review 1 1/2 hrs.
status of non 'D4I-2 open items a.
b.
status of IE review 1
c.
new W reactor generic items d.
implTeations of issuance of a low-power licensee status of Sequoyah seismic reevaluation e.
f.
status report on UHI brief report on the status of the McGuire g.
/
review LUNCH I hr.
- 12:30 p.m.
DISCUSSION OF SEQUOYAH/ ICE CDCENSER - UHI PLANTS (CONT'D) 1:30 p.m.
(4) Status of the NRC Staff review of the Sequoyah response to the Lessons tar.rned I hr.
Task Force recommendations Ib-(5) Status of the NRC review of the Sequoyah 30 mins.
response to the Bulletins and Orders (6) NRC response to ACRS 'IMI-2 Implications Subcomittee questions (See attached letter, 7
dated October 12, 1979, Savio & Major to 30 mins.
Vassallo)
(7) Updated report on hydrogen control in ice 30 mins.
condenser containments - F. Rowsome
'IMI-2 IMPLICATIONS AS 'IHEY REIAE TO DIABLO CANYON 4:00 p.m.
2 hrs.
AND SIMIIAR PIANTS NRC response to ACRS 'IMI-2 Implications StLWittee questions (see attached letter, dated October 12, 1979, cgQ)&G6Ihl
- (Y 'to / >
Savio & Major to Vassallo)
(,)Y y
Al:4x'nuutne C
- 1 ADDITIONAI. ITENS 6:00 p.m.
Staff review of emergency procedures - brief (1) status report Operator Actions - actions an operator may take (2) under emergency conditions which would further degrade the condition of the plant Pressurizer heaters - potential failure modes for (3) pressurizer heaters and associated consequences f
Shutdown heat removal capacity available to operator I
(4) following an accident for various degraded plant conditions 8:00 p.m.
AIUOURN
- i
- !*g#f taltTED STATES NUCLEAR REGULATORY COMMISSION l
ADVl80RY COMMITTEE ON REACTOR SAFEOUARDS siaan m orow o.c.sesas October 12, 1979 D. B. Vassallo, Acting Director, Division of Project Management ACRS QUESTICNS REIATING '!O CERMIN EAR-TEM CL PLANT EULTICT:
OMfCN, EDMER, SEQUORE, AND MCX;UIRE)
Attached for your use are the questions dich have been p
-ittee meetings for near-term operating liennaes on the Diablo Canyon, Eisumer, Md;uire, and sequoyah nuclear plants.
Please let me know een your ' staff can be prepared to discuss these issues ao that we can proceed with ACRS consideration of these projects.
s A written response is preferred but the 9_hittee would be prepared to i
discuss these matters orally to facilitate progress on these matters.
S Richard Savio Staff Engineer ACRS Staff Richard K.
jor Reactor Engineer ACRS Staff
Attachment:
As stated
)
(5
/
DUPLICATE DOCUMENT Entire document previously entered into system under:
b
!O[b ANO No. of pages: _
]8 K(-. g/L 57G AL gAC 37x-.cc The following discussion has been prepared by the staff to address question III.1 of the "ACRS Questions Relating to Certain Near-Term OL Plants (Diablo Canyon, Zimmer, Sequoyah and McGuire)" which were transmitted by Messrs. SaWio and Major s
letter of October 12, 1979 to D. B. Vassallo, Acting Director of the Division of Project Management. The question speci(ically request, that the staff discuss the special features of the ice condenser /UHI plants which would require consider-ation in the light of the TMI-2 accident.
In the following discussion we will specifically address the post accident containment hydrogen control capability for ice condenser plants, and primary system blockage by non-condensible gases.
==
Introduction:==
The accident at TMI-2 wa$ the result of a combination of equipment failures, pro-cedural errors, operator errors and design deficiencies. The accident resulted in core damage including hydrogen generation due to metal-water reaction of the fuel element cladding which greatly exceeded the current design basis.
We have reviewed the Sequoyah and McGuire Nuclear Plant designs and have detennined that the only plant design features which appear to be unique to these plants are the ice condenser containment and the upper head injection systems.
We have examinad these systems with respect to the lessons learned from the TMI-2 accident and have reached the conclusions detailed below.
Ice Condenser Containment Desian The ice condenser is a passive pressure suppression system requiring no operator action to be effective during a design basis event.
Use of the ice condenser de-sign has resulted in a small containment building with a low internal design pres-sure. When a beyond-design-basis event like TMI-2 is considered for the ice con-denser design the above design features lead to questions regarding the ability of the containment building to withstand the release and combustion of large amounts of hydrogen in the building, i.e., post accident hydrogen control capability.
~
_g.
We have examined this aspect of the ice condenser containment design and in our judgment we feel that the structural integrity of;)he containment building cannot be assured if hydrogen in the amount generated during the TMI-2 accident were to be released and burned in an ice condenser containmant building.
In Chapter 3, particularly section 3.3, of NUREG-8585, "TMI-2 Lessons Learned Task Force Final Report," there is a discussion of the need for and feasibility of hydrogen control features in all LWRs that would go beyond the current design bases specified in the NRC regulations.
In recomendation 10 of that report, the Task Force recommended to the Director of Nuclear Reactor Regulation that the Comission give notice of intent to conduct rulemaking relating to the consider-ation of design features to mitigate degraded core and core melt accidents, in particular, systems for preventing the uncontrolled combustion of hydrogen that could be produced in such accidents. The director of NRR has asked the ACRS to review and coment on the recomendations in flVREG-0585, and the Office is concur-rently reviewing ^he recomendations in context with those o.f the President's j
Comission on the accident at Three Mile Island and others.
The results of that review will be presented to the Comission for decision.
A decision on whether and how to proceed with the proposed rulemaking is not expected to be made by the Comission for several months.
The staff proposes to defer impositon of requirements on hydrogen control beyond present requirements pending a decision by the Comission on the proposed rule making.
Of primary importance will be the new metal-water reaction rate to be assumed.
In the course of rule making, we would expect to consider the feasibility l
of various alternatives, such as inerting, filtered venting, and controlled burn-ing as hydrogen gas treatment systems.
. We believe deferral of further action at this time relative to hydrgen control is justified. The design of the Sequoyah and McGu{re Nuclear Plants will be re-quired to accomodate the accident prevention measures currently being introduced by the Lessons Learned and Bulletins and,0rders task forces, those recommended by the President's Commission, and others. These measures include changes in safety equipment design, operator training, accident response, and diagnostic instrumen-tation to reduce the probability of future accidents which, like TMI-2, might ex-ceed the current design basis and produce large amounts of hydrogen.
On this basis, we propose that for an interim period, until rule making can be conducted, the licensing process for. these plants can continue to be conducted in accordance with the current regulations and guides for the design and installation of post-accident containment combustible gas control system, i.e., paragraph 50.44 to 10 CFR Part 50 and Regulatory Guide 1.7.
The post-accident Combustible Gas Control System in ice condenser containments consists of redundant safety grade internal recombimers and containment atmos-phere mixing systems.
The recombiners are 100 SCFM Westinghouse thermal hydrogen recombiners.
Each train of the mixing systems is comprised of a Return Air Fan System to promote recirculation of containment atmosphere from the upper compart-ment to the lower compartment and a hydrogen skimer system to prevent the ac-cumulation of localized concentrations of hydrogen in the containment subcompart-ments and reactor building dome. These systems were originally designed to ac-comodate the release of_ hydrogen into the containment from an assumed 5% metal-water reaction in the reactor core. Assuming that these systems are adequate to assure complete mixing in the event of an accident resulting in greater than a 5% metM, mer reaction, an ice condenser plant can accomodate the hydrogen pro-duced by somewhat greater than a 15% metal-water reaction without exceeding the
A.
.' design basis of the post-accident combustible gas control system, i.e., the abil-ity to prevent the hydrogen concentration in the cgntainment building from exceed-ing the lower limit of flamability for hydrogen in air (4 volume percent). Thus, there is capability interent in the design to accommodate a beyond-design-basis event of severity somewhat less than the TMI-2 event.
Follow-on studies to the Reactor Safety Study (WASH 1400) have been completed for an ice condenser plant with a free standing steel containment building having a design pressure of 12 psig.
Failure of the containment building was predicted to occur at 27 ! 3 psig.
For core melt accidents (assuming 100% metal-water reaction) the Reactor Safety Study concluded that containment failure as a result of hydro-gen c'ombustion would be a virtual certainty.
If the amount of hydrogen is instead assumed to be that produced by a near-melt accident (such as TMI-2) and a 50%
metal-water reaction is considered, the likelihood of containment failure due to hydrogen combustion is not significantly changed. The Reactor Safety Study has indicated that failure of the containment from hydrogen combustion would not be the expected consequence for plants utilizing a large dry containment design.
If the outcome of the proposed rule making is such that mitigat1ng measures are 1
required for beyond-design-basis events resulting in severe core damage, but not substantial melting, it appears that plants using ice condenser containment de-signs would have to provide a means of preventing or controlling combustion in the l
containment.
This might be accomplished by pre-or post-accident systems for in-erting of the containment building or possibly by the use of distributed ignition sources in the containment building.
Any attempt to impose immediate requirements for such systems at this time would,in our judgment,be premature.
Evaluation of the feasibility and merits of systems of this type should be embodied within the rule making process.
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5-Should the outcome of the rule making process requ$re mitigating measures for core melt accidents, filtered overpressure relief systems for ice condenser con-5 tainments and dry containments may become a requirement.
Again, the imposition of imediate requirements for such a system would also be premature.
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Non-Combustible Gases in the Reactor Coolant System All plants using the ice condenser containment design, except the D. C. Cook, Units 1 and 2, have been equipped with an upper head injectio.n (UHI) system.
The purpose of the UHI system is to provide enhanNd reactor cooling capability duringalargebreakLOCA,forplantsw(ichwillbeexpectedtohavealowback pressure (containment pressure) during the core reflooding transient.
The UHI system utilizes a high pressure 1400 cubic ft. nitrogen accumulator tank as a driver for a water filled accumulator tank.
Redundant (safety grade) stop valves are used in the discharge piping to prevent the :ccidental release of the nitrogen gas into the reactor coolant system.
During the accident at 19lI-2 a large amount of gaseous hydrogen was produced by metal-water reaction of the fuel cladding during periods of high fuel temperature.
Natural circulation cooling was seriously impaired by the non-condensible gas.
The Commission has since required that all plants design and install high point vents in the reactor coolant systems to allow the release of non-condensible gases.
This design feature has been required to mitigate the effects of a beyond-design-basis event like TMI-2.
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Within the framework of our current licensing review criteria, i.e., single failure, we would not assume the release of the UHI nitrogen accumulator into the RCS.
We may, in the future, require that for systems with a high safety impact, reliability assessments be made on the basis of a probabilistic approach in addition to the single failure assumption. This issue is discussed more fully in Recommendation 8 of the TMI-2 Lessons Learned Task Force Final Report, NUREG 0585.
In the interim, it would appear prudent to consider the effects of nitrogen addition of'both UHI stop valves when setting the RCS vent system flow capacities and in operator train-ing ud emergency procedures.
Preliminary calculations by the staff indicate that
I.
2-a vent system designed to accommodate the non-condensible gas (hydrogen) from metal-water reaction of the extent that occurred at TMI-2 (35.to 50%) would prob-ably be sufficient to provide a relief path for inadvertent release of the UHI sys*.em nitrogen accumulator into the RCS. We will pursue this in our reviews of the Sequoyah and McGuire plants, specifically, during our reviews of their RCS vent systen designs and their emergency procedures.
The small break analyses required by Bulletin 79-06A and Short Term Lessons Learned Requirement 2.1.9 have not been perfomed for UHI plants.
We will require that these analyses be performed using the UHI ECCS evaluation model, in combina-tion with best estimate input, prior to licensing of the Sequoyah and McGuire nuclear plants.
The Semi-Scale Test program includes a series of tests for an ECC system using upper head injection. These tests are scheduled for FY 81.
We will require that Sequoyah and McGuire perform pretest predictions of a small break LOCA test using the UHI ECCS evaluation model.
Following the completion of the semi-scale tests we will compare the results with the applicants' pre-test predictions.
We regard these analyses as confirmatory in nature and will not condition the licensing of Sequoyah and McGuire upon the completion of the UHI semi-scale experiments and the pretest prediction comparisons.
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UPPER HEAD INJECTION NITROGEN PREVENTION MEASURES DURING INJECTION l
9 MINIMIZE N2 FOLLOWING INJECTION BY ISOLATION SYSTEM l
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STATUS OF SEQUOYAH NON-TMi-2 OPEN ITEMS
(* ltems Listed in SER)
ITEM #
TASK DESCRIPTION STATUS ITEM =
TASK DESCRIPTION STATUS 1
21 INSTRUMENT TRIP SET POINTS V*
EFFECT OF POWER TRANSIENTS 2
ATWS 22 (01( UNTIL IST REFUELING) 23 SEISMIC DESIGN - SD EGPT.
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3 OVERPRESSU,RE PROTECTION 4
4 SIS BYPASS V*
24 APP. J. E XEMPTIONS V
5 LOCA ANALYSIS 25 SG & PZR SUPPORT DESIGN V*
G IST OF PUMPS & VALVES 2G APP. G & H EXEMPTIONS V"
7 BTP RSB 5-1 27 S. G. TUBE (S) s/
- 8 RHR SINGLE FAILURE V'
28 PLANT TRIP TEST V*
9 UH1 PREOP TEST V*
29 DIESEL & REMOTE SD TESTING V"
10 CONTAINMENT SUMP TEST 30 FIRE PROTECTION V*
31 SEISMIC DESIGN - %T R V
11 BORON DILUTION V*
e 12 LONG-TERM SLB EFFECTS V*
32 LOOSE PARTS MONITOR V*
13 ROD DROP TRANSIENT V
33 SEISMIC DESIGN - %/6 V
14 CONTRGL SPIDER V
34 FOUNDATION ENGINEERING V
15 GRif, STRArs V
35 SECURITY PLAN V
16 GUIDE TUBE WEAR V
36 CLOSURE HEAD FLAW V
17 ENVIRONMENTAL QUAL.
37 OPERATOR TR AINING 38 BY-PASS LEAKAGE 18 SEISMIC QUAL.
19 RESPONSE TIME TESTING V*
39 SECONDARY WATER CHEMISTRY V
l 20 ISOLATION VALVE INTERLOCKS y*
40 AFW RUNOUT FLOW NOTE: DOES NOT INCLUDE IE ITEMS THAT HAVE BEEN FOLLOWED BY NRR STAFF.
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IDBRIFY BNIR0ffEff RESULTIfG FROM EXittiw PERFORMANCE OF NONSEISMICFLUIDSYSTEE o
IDENTIFY EXPELlw PERFORWICE OF NON SEISMIC EDUIPME!#
GIVEN SEISMIC #1D CONSEWBRIAL BNIR0fiEff o
IDBEIFY SYSTBE IfPACT OF NON SEISMIC EWIREff #0 SYSTBE PERf0W#lCE
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